Computer Programs
DLC-0031 ZZ-FEWG1, ZZ-GEDT1/4
ZZ DLC-31, 37 Neutron-Group, 21 Gamma-Group Coupled Group Constants Library from ENDF/B
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROBLEM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
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To submit a request, click below on the link of the version you wish to order. Rules for end-users are
available here.
| Program name | Package id | Status | Status date |
|---|---|---|---|
| ZZ-GEDT1/400 | DLC-0031/02 | Tested | 01-JUL-1979 |
| ZZ-GPDT1/402B | DLC-0031/03 | Tested | 01-SEP-1979 |
| ZZ-NEDT1/401 | DLC-0031/04 | Tested | 01-JUL-1979 |
| ZZ-FEWG1/H | DLC-0031/06 | Tested | 27-SEP-1988 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| DLC-0031/02 | IBM 360 series | IBM 360 series |
| DLC-0031/03 | IBM 360 series | IBM 360 series |
| DLC-0031/04 | IBM 360 series | |
| DLC-0031/06 | IBM 3033 | DEC VAX 11/780 |
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3. DESCRIPTION OF PROBLEM OR FUNCTION
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FEWG1/H
=======
FORMAT: ANISN
NUMBER OF GROUPS: 37-neutron, 21-gamma. Energy range 1.0E-11 to 19.64 MeV for neutrons and 0.01 to 14.0 MeV for gamma rays.
NUCLIDES: H, H-3, Li-6, Li-7, Be-9, B-10, B-11, C-12, N, O, F, Na, Mg, Al, Si, P-31, S-32, Cl, Ar, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Mo, Ba, Gd, Ta-181, W-182, W-183, W-184, W-186,Pb, U-235, U-238, Pu-238, Pu-240.
ORIGIN: DNA Working Cross Section Library and the Evaluated Nuclear Data File (ENDF/B)
WEIGHTING SPECTRUM: 1/E and 300K Maxwellian weighting spectra were used for neutrons.
GEDT1/400
=========
FORMAT: Gamma-ray cross section library in SAM-CE format.
NUMBER OF GROUPS: Data are given for the energy range from 1 keV to 100 MeV.
NUCLIDES: Data are included for materials with atomic numbers 1-83, 86, 90, 92, 94, and for some naturally occuring elements.
ORIGIN: DLC-0007
WEIGHTING SPECTRUM: --
GPDT1/402B
==========
FORMAT: Gamma-ray production data in SAM-CE format.
NUMBER OF GROUPS: --
NUCLIDES: Data are included for H-1, N, O, Al, Si, Ca, Pb, Li-7, B-10, C-12, Mg, Fe, U-235, U-238, Cl, W-182, W-183, W-184, W-186.
ORIGIN: ENDF/B-IV
WEIGHTING SPECTRUM:
NEDT1/401
=========
FORMAT: Nuclear element data in SAM-CE format.
NUMBER OF GROUPS:
NUCLIDES: Data are included for H-1, H-3, He-4, Li-7, B-10, B-11, C-12, N, O, Na, Mg, Al, Si, Ca, Fe, Cu, Pb, W-182, W-183, W-184, W-186, U-235, U-238, Pu-239, Pu-240.
ORIGIN: ENDF/B-IV
WEIGHTING SPECTRUM:
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
ZZ-FEWG1: Coupled 37-neutron, 21-gamma group, P3, multigroup library in ANISN format, based on data from the DNA Working Cross Section Library and the Evaluated Nuclear Data File. Cross sections for the following materials are included: H, H-3, Li-6, Li-7,Be-9, B-10, B-11, C-12, N, O, F, Na, Mg, Al, Si, P-31, S-32, Cl, Ar, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Mo, Ba, Gd, Ta-181, W-182, W-183, W-184, W-186, Pb, U-235, U-238, Pu-238, Pu-240. The energy range is 1.0E-11 to 19.64 MeV for neutrons and 0.01 to 14.0 MeV for gamma rays. 1/E and 300K Maxwellian weighting spectra were used for neutrons.
ZZ-GEDT1/400: Gamma-ray cross section library in SAM-CE format. Data are included for materials with atomic numbers 1-83, 86, 90, 92, 94, and for some naturally occuring elements. Data are given for the energy range from 1 keV to 100 MeV.
ZZ-GPDT1/402B: Gamma-ray production data in SAM-CE format. Data are included for H-1, N, O, Al, Si, Ca, Pb, Li-7, B-10, C-12, Mg, Fe, U-235, U-238, Cl, W-182, W-183, W-184, W-186.
ZZ-NEDT1/401: Nuclear element data in SAM-CE format. Data are included for H-1, H-3, He-4, Li-7, B-10, B-11, C-12, N, O, Na, Mg, Al, Si, Ca, Fe, Cu, Pb, W-182, W-183, W-184, W-186, U-235, U-238, Pu-239, Pu-240.
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
FEWG1/H
=======
FORMAT: ANISN
NUMBER OF GROUPS: 37-neutron, 21-gamma. Energy range 1.0E-11 to 19.64 MeV for neutrons and 0.01 to 14.0 MeV for gamma rays.
NUCLIDES: H, H-3, Li-6, Li-7, Be-9, B-10, B-11, C-12, N, O, F, Na, Mg, Al, Si, P-31, S-32, Cl, Ar, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Mo, Ba, Gd, Ta-181, W-182, W-183, W-184, W-186,Pb, U-235, U-238, Pu-238, Pu-240.
ORIGIN: DNA Working Cross Section Library and the Evaluated Nuclear Data File (ENDF/B)
WEIGHTING SPECTRUM: 1/E and 300K Maxwellian weighting spectra were used for neutrons.
GEDT1/400
=========
FORMAT: Gamma-ray cross section library in SAM-CE format.
NUMBER OF GROUPS: Data are given for the energy range from 1 keV to 100 MeV.
NUCLIDES: Data are included for materials with atomic numbers 1-83, 86, 90, 92, 94, and for some naturally occuring elements.
ORIGIN: DLC-0007
WEIGHTING SPECTRUM: --
GPDT1/402B
==========
FORMAT: Gamma-ray production data in SAM-CE format.
NUMBER OF GROUPS: --
NUCLIDES: Data are included for H-1, N, O, Al, Si, Ca, Pb, Li-7, B-10, C-12, Mg, Fe, U-235, U-238, Cl, W-182, W-183, W-184, W-186.
ORIGIN: ENDF/B-IV
WEIGHTING SPECTRUM:
NEDT1/401
=========
FORMAT: Nuclear element data in SAM-CE format.
NUMBER OF GROUPS:
NUCLIDES: Data are included for H-1, H-3, He-4, Li-7, B-10, B-11, C-12, N, O, Na, Mg, Al, Si, Ca, Fe, Cu, Pb, W-182, W-183, W-184, W-186, U-235, U-238, Pu-239, Pu-240.
ORIGIN: ENDF/B-IV
WEIGHTING SPECTRUM:
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
ZZ-FEWG1: Coupled 37-neutron, 21-gamma group, P3, multigroup library in ANISN format, based on data from the DNA Working Cross Section Library and the Evaluated Nuclear Data File. Cross sections for the following materials are included: H, H-3, Li-6, Li-7,Be-9, B-10, B-11, C-12, N, O, F, Na, Mg, Al, Si, P-31, S-32, Cl, Ar, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Mo, Ba, Gd, Ta-181, W-182, W-183, W-184, W-186, Pb, U-235, U-238, Pu-238, Pu-240. The energy range is 1.0E-11 to 19.64 MeV for neutrons and 0.01 to 14.0 MeV for gamma rays. 1/E and 300K Maxwellian weighting spectra were used for neutrons.
ZZ-GEDT1/400: Gamma-ray cross section library in SAM-CE format. Data are included for materials with atomic numbers 1-83, 86, 90, 92, 94, and for some naturally occuring elements. Data are given for the energy range from 1 keV to 100 MeV.
ZZ-GPDT1/402B: Gamma-ray production data in SAM-CE format. Data are included for H-1, N, O, Al, Si, Ca, Pb, Li-7, B-10, C-12, Mg, Fe, U-235, U-238, Cl, W-182, W-183, W-184, W-186.
ZZ-NEDT1/401: Nuclear element data in SAM-CE format. Data are included for H-1, H-3, He-4, Li-7, B-10, B-11, C-12, N, O, Na, Mg, Al, Si, Ca, Fe, Cu, Pb, W-182, W-183, W-184, W-186, U-235, U-238, Pu-239, Pu-240.
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DLC-0031/06
The BCD to binary conversion program was run by NEA-DB against the data library on a VAX-11/780 computer in 3min 15sec.[ top ]
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| Package ID | Status date | Status |
|---|---|---|
| DLC-0031/02 | 01-JUL-1979 | Tested at NEADB |
| DLC-0031/03 | 01-SEP-1979 | Tested at NEADB |
| DLC-0031/04 | 01-JUL-1979 | Tested at NEADB |
| DLC-0031/06 | 27-SEP-1988 | Tested at NEADB |
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10. REFERENCES
- N.M. Greene, et al.,
"AMPX: A Modular Code System for Generating Coupled Multigroup
Neutron-Gamma Libraries from ENDF/B",
ORNL-TM-3706 (March 1976).
- D. Garber, C. Dunford, and S. Pearlstein,
"Data Formats and Procedures for the Evaluated Nuclear Data File,
ENDF",
BNL-NCS-50496 (ENDF 102) Revised Oct. 1975.
- D.E. Bartine, E.M. Oblow, F.R. Mynatt,
"Radiation-Transport Cross-Section Sensitivity Analysis -
A General Approach Illustrated for a Thermonuclear Source in Air", Nuc. Sci. Eng. 55, 147 (1974).
- W.W. Engle, Jr.,
"A User'S Manual for ANISN, A One-Dimensional Discrete Ordinates
Transport Code with Anisotropic Scattering",
K-1693, Computing Technology Center (1976).
- J.J. Ritts, M. Solomito, P.N. Stevens,
"The Calculation of Neutron-Induced Physical Doses in Human
Tissues",
ORNL-TM-2991 (1969).
- M.A. Abdou, C.W. Maynard, R.Q. Wright,
"MACK: A Computer Program to Calculate Neutron Energy Release
Parameters (fluence-to-Kerma Factors) and Multigroup Neutron Reac- tion Cross Sections From Nuclear Data in ENDF Format",
ORNL-TM-3994 (July 1973).
- "Protection Against Neutron Radiation",
NCRP-38, National Council on Radiation Protection and Measurements Washington, D.C. (1971).
- H.C. Claiborne, D.K. Trubey,
"Dose Rates in a Slab Phantom from Monoenergetic Gamma Rays",
Nucl. Appl. Tech. 8, 450 (1970).
- W.S. Snyder and C. Neufeld,
Radiation Res., 6, 1, 67 (1957).
- B.J. Henderson,
"Conversion of Neutron or Gamma Ray Flux to Absorbed Dose Rate",
XDC 59-8-179 (August 14, 1959).
- E.A. Straker and M.L. Gritzner,
"Neutron and Secondary Gamma-Ray Transport in Infinite Homogeneous Air",
ORNL-TM-4464 (1969).
- J.V. Pace, III, D.E. Bartine, F.R. Mynatt,
"Neutron and Secondary-Gamma-Ray Transport Calculations for 14-MeV and Fission Neutron Sources in Air-Over-Ground and Air-Over-Sea-
water Geometries",
ORNL-TM-4841 (August 1975).
- P.G. Fisher, W.S. Knapp,
"Aids for the Study of Electromagnetic Blackout",
70TMP-12, General Electric - TEMP, Santa Barbara, Cal., p. 2.7
(July 1970).
- N.M. Greene, et al.,
"AMPX: A Modular Code System for Generating Coupled Multigroup
Neutron-Gamma Libraries from ENDF/B",
ORNL-TM-3706 (March 1976).
- D. Garber, C. Dunford, and S. Pearlstein,
"Data Formats and Procedures for the Evaluated Nuclear Data File,
ENDF",
BNL-NCS-50496 (ENDF 102) Revised Oct. 1975.
- D.E. Bartine, E.M. Oblow, F.R. Mynatt,
"Radiation-Transport Cross-Section Sensitivity Analysis -
A General Approach Illustrated for a Thermonuclear Source in Air", Nuc. Sci. Eng. 55, 147 (1974).
- W.W. Engle, Jr.,
"A User'S Manual for ANISN, A One-Dimensional Discrete Ordinates
Transport Code with Anisotropic Scattering",
K-1693, Computing Technology Center (1976).
- J.J. Ritts, M. Solomito, P.N. Stevens,
"The Calculation of Neutron-Induced Physical Doses in Human
Tissues",
ORNL-TM-2991 (1969).
- M.A. Abdou, C.W. Maynard, R.Q. Wright,
"MACK: A Computer Program to Calculate Neutron Energy Release
Parameters (fluence-to-Kerma Factors) and Multigroup Neutron Reac- tion Cross Sections From Nuclear Data in ENDF Format",
ORNL-TM-3994 (July 1973).
- "Protection Against Neutron Radiation",
NCRP-38, National Council on Radiation Protection and Measurements Washington, D.C. (1971).
- H.C. Claiborne, D.K. Trubey,
"Dose Rates in a Slab Phantom from Monoenergetic Gamma Rays",
Nucl. Appl. Tech. 8, 450 (1970).
- W.S. Snyder and C. Neufeld,
Radiation Res., 6, 1, 67 (1957).
- B.J. Henderson,
"Conversion of Neutron or Gamma Ray Flux to Absorbed Dose Rate",
XDC 59-8-179 (August 14, 1959).
- E.A. Straker and M.L. Gritzner,
"Neutron and Secondary Gamma-Ray Transport in Infinite Homogeneous Air",
ORNL-TM-4464 (1969).
- J.V. Pace, III, D.E. Bartine, F.R. Mynatt,
"Neutron and Secondary-Gamma-Ray Transport Calculations for 14-MeV and Fission Neutron Sources in Air-Over-Ground and Air-Over-Sea-
water Geometries",
ORNL-TM-4841 (August 1975).
- P.G. Fisher, W.S. Knapp,
"Aids for the Study of Electromagnetic Blackout",
70TMP-12, General Electric - TEMP, Santa Barbara, Cal., p. 2.7
(July 1970).
DLC-0031/02, included references:
- D.E. Bartine et al.:DNA Processed Libraries for SAM-CE
(1976)
DLC-0031/03, included references:
- D.E. Bartine et al.:DNA Processed Libraries for SAM-CE
(1976)
DLC-0031/04, included references:
- D.E. Bartine et al.:DNA Processed Libraries for SAM-CE
(1976)
DLC-0031/06, included references:
- D.E. Bartine, J.R. Knight, J.V. Pace and R.W. Roussin:Production and Testing of the DNA Few-Group Coupled Neutron-Gamma
Cross-Section Library. ORNL/TM-4840 (March 1977)
- COMAND - AMPX Module to Collapse ANISN Cross Section Libraries.
Informal Notes
- L.R. Williams:
LIBGEN - A Special Version of the ANISN Library Generation Routine
Informal Notes (1972)
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| Package ID | Computer language |
|---|---|
| DLC-0031/02 | FORTRAN-IV |
| DLC-0031/03 | FORTRAN-IV |
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DLC-0031/02
| File name | File description | Records |
|---|---|---|
| DLC0031_02.001 | INFORMATION | 4 |
| DLC0031_02.002 | Z RANGE 1-10 | 1151 |
| DLC0031_02.003 | Z RANGE 11-20 | 801 |
| DLC0031_02.004 | Z RANGE 21-30 | 713 |
| DLC0031_02.005 | Z RANGE 31-40 | 749 |
| DLC0031_02.006 | Z RANGE 41-50 | 767 |
| DLC0031_02.007 | Z RANGE 51-60 | 797 |
| DLC0031_02.008 | Z RANGE 61-70 | 838 |
| DLC0031_02.009 | Z RANGE 71-80 | 1254 |
| DLC0031_02.010 | Z RANGE 81,82,83,86,90,92,94 | 907 |
| DLC0031_02.011 | BCDEAN SOURCE | 271 |
| DLC0031_02.012 | BCDEAN DD CARDS | 10 |
| DLC0031_02.013 | INPUT | 22 |
| DLC0031_02.014 | OUTPUT | 128 |
DLC-0031/03
| File name | File description | Records |
|---|---|---|
| DLC0031_03.001 | INFORMATIONS | 4 |
| DLC0031_03.002 | SOURCE PROGRAM (F4,EBCDIC) | 271 |
| DLC0031_03.003 | JCL | 10 |
| DLC0031_03.004 | SAMPLE OUTPUT | 8 |
| DLC0031_03.005 | SAMPLE INPUT DATA | 2 |
| DLC0031_03.006 | LIBRARY (A1-27) | 805 |
| DLC0031_03.007 | LIBRARY (B-10) | 133 |
| DLC0031_03.008 | LIBRARY (C-12) | 164 |
| DLC0031_03.009 | LIBRARY (N-14) | 843 |
| DLC0031_03.010 | LIBRARY ( MG ) | 358 |
| DLC0031_03.011 | LIBRARY (U-238) | 98 |
| DLC0031_03.012 | LIBRARY (O-16) | 763 |
| DLC0031_03.013 | LIBRARY ( SI ) | 2673 |
| DLC0031_03.014 | LIBRARY ( CL ) | 743 |
| DLC0031_03.015 | LIBRARY (W-182) | 562 |
| DLC0031_03.016 | LIBRARY (W-183) | 608 |
| DLC0031_03.017 | LIBRARY (W-184) | 500 |
| DLC0031_03.018 | LIBRARY (W-186) | 467 |
| DLC0031_03.019 | LIBRARY ( H-1 ) | 8 |
| DLC0031_03.020 | LIBRARY (PB-NATURAL) | 220 |
| DLC0031_03.021 | LIBRARY ( CA ) | 870 |
| DLC0031_03.022 | LIBRARY ( LI-7 ) | 26 |
| DLC0031_03.023 | LIBRARY ( FE ) | 1511 |
| DLC0031_03.024 | LIBRARY (U-235) | 2060 |
DLC-0031/04
| File name | File description | Records |
|---|---|---|
| DLC0031_04.001 | INFORMATION | 4 |
| DLC0031_04.002 | LIBRARY (NITROGEN) | 2377 |
| DLC0031_04.003 | LIBRARY (HYDROGEN) | 94 |
| DLC0031_04.004 | LIBRARY (SILICON) | 6811 |
| DLC0031_04.005 | LIBRARY (OXYGEN) | 3413 |
| DLC0031_04.006 | LIBRARY (ALMINUM) | 2185 |
| DLC0031_04.007 | LIBRARY (LEAD,NATURAL) | 2502 |
| DLC0031_04.008 | LIBRARY (SODIUM) | 3548 |
| DLC0031_04.009 | LIBRARY (H-3) | 161 |
| DLC0031_04.010 | LIBRARY (HE-4) | 252 |
| DLC0031_04.011 | LIBRARY (C-12) | 1526 |
| DLC0031_04.012 | LIBRARY (CA) | 3081 |
| DLC0031_04.013 | LIBRARY (FE) | 7121 |
| DLC0031_04.014 | LIBRARY (CU) | 1169 |
| DLC0031_04.015 | LIBRARY (W-182) | 1629 |
| DLC0031_04.016 | LIBRARY (W-186) | 1378 |
| DLC0031_04.017 | LIBRARY (U-238) | 1195 |
| DLC0031_04.018 | LIBRARY (LI-7) | 415 |
| DLC0031_04.019 | LIBRARY (B-10) | 1293 |
| DLC0031_04.020 | LIBRARY (B-11) | 524 |
| DLC0031_04.021 | LIBRARY (MG) | 3815 |
| DLC0031_04.022 | LIBRARY (K) | 1370 |
| DLC0031_04.023 | LIBRARY (W-183) | 1315 |
| DLC0031_04.024 | LIBRARY (W-184) | 1261 |
| DLC0031_04.025 | LIBRARY (U-235) | 2136 |
| DLC0031_04.026 | LIBRARY (PU-239) | 1384 |
| DLC0031_04.027 | LIBRARY (PU-240) | 1731 |
| DLC0031_04.028 | BCDEAN SOURCE | 271 |
| DLC0031_04.029 | BCDEAN DD CARDS | 10 |
| DLC0031_04.030 | BCDEAN DATA | 2 |
| DLC0031_04.031 | BCDEAN OUTPUT | 8 |
DLC-0031/06
| File name | File description | Records |
|---|---|---|
| DLC0031_06.001 | INFORMATION FILE | 91 |
| DLC0031_06.002 | FEWG1-85 CROSS SECTION LIBRARY | 17470 |
| DLC0031_06.003 | BCBN BCD TO BINARY CONVERSION PROGRAM | 45 |
| DLC0031_06.004 | ORIGINAL JCL FOR BCD TO BINARY RUN ON IBM | 19 |
| DLC0031_06.005 | COMMAND PROC.FOR BCD TO BINARY RUN ON VAX | 9 |
| DLC0031_06.006 | OUTPUT FROM BCD TO BINARY CONVERSION RUN | 174 |
Keywords: coupled neutron gamma cross sections, cross sections, data, gamma radiation, group constants, neutron transport theory, photon transport, shielding.